1. Field of the Invention
This invention pertains in general to the storage of spent nuclear fuel and more particularly to a method of calculating the burnup credit for spent nuclear fuel rods.
2. Description of the Prior Art
Transportation and shipment of spent nuclear fuel for eventual disposal is regulated by the Nuclear Regulatory Commission (NRC) through the requirements of Title 10 of the Code of Federal Regulations, Part 71. To meet the requirements of 10 CFR §71, transportation casks must be designed to ensure criticality safety. The safety analyses for these transportation casks are presently based on the assumption that the fuel assemblies are unirradiated, i.e., the fissile content is the same as the as-manufactured assembly. This assumption is conservative in the spent nuclear fuel case, as the fissile isotopes have been burned up as a result of the use of the fuel assembly in a reactor and, therefore, the fissile isotope content of the assembly is much lower than the as-manufactured content.
The capacity of transportation casks can be severely limited by the “fresh fuel” assumption, as larger criticality safety margins exist in the spent fuel assembly case. If credit could be taken for the burnup of the assemblies, cost savings in the transportation of spent fuel assemblies would result. In the development of canisters in support of dry storage of spent nuclear fuel, approval has been sought for a burnup credit methodology in support of package loading. The burnup credit methodology will rely on a combination of calculated burnup using reactor records, and burnup verification measurements to verify reactor records.
With increasing emphasis on issues related to the shipment of fuel for eventual disposition, burnup verification measurements and methodology are assuming a role of greater importance. The “fresh fuel” assumption results in very conservative designs for spent fuel racks, shipping canisters and waste repository storage. These overly conservative designs result in increased costs for the storage and shipping of spent nuclear fuel.
In order to take advantage of burnup credit for spent nuclear fuel, a method must be in place to reliably verify the fissionable content of fuel assemblies to ensure that criticality safety limits are not exceeded. Typical burnup verification methods require measurements of fuel assemblies to confirm reactor records of initial enrichment, burnup and decay time. These measurements rely on determining the neutron flux and, in some cases, the gamma dose in the vicinity of the fuel centerline.
The measurement technology in presently available commercial systems relies on 235U fission chambers to measure the neutron specific activity and either gamma ionization chambers, gamma scintillation detectors or solid-state semiconductor detectors (high-purity germanium HPGe), to detect gamma rays. Fission chambers and gamma ionization chambers are rather large gas-filled detectors. NaI(Tl) scintillation gamma ray detectors are typically large and require a photomultiplier tube and gamma ray shielding for operation in a spent fuel environment. HPGe gamma ray detectors require a liquid nitrogen cryogenic system or an electronic cooling system, since they are not capable of operation as high-resolution gamma ray detectors at higher temperatures. These detectors are sensitive to environmental factors such as temperature and the intense, mixed gamma ray and neutron field. For example, fission chambers are sensitive to gamma ray background, and gamma ionization chambers, NaI(Tl) detectors and HPGe detectors are all sensitive to neutron-induced background.
During the course of the use of a fuel assembly in a reactor core, higher actinides are produced by a chain of neutron captures followed by beta decay. In uranium fuel, the higher-actinide buildup chain originates with the 238U present in the fuel. Many of the higher actinides decay by spontaneous fission, a process which is accompanied by the emission of neutrons associated with fission. A secondary source of neutrons exists in oxide fuels where neutrons can be produced via the action of energetic alpha particles (primarily from the decay of higher actinides) on the 18O isotope of oxygen. It has been demonstrated by many workers that the neutron specific activity of spent fuel is related to burnup. Detailed mathematical relationships between the neutron emission rate and burnup have also been inferred using measurements on spent fuel assemblies. The functional form of this relationship is that the neutron emission rate is a function of assembly burnup raised to a power.
Variables that affect the neutron emission rate include fuel type, initial enrichment, power history and decay time since discharge of the fuel assembly from the reactor.
Although spontaneously fissioning plutonium isotopes and plutonium alpha emitters are the dominant source of neutrons during the first fuel operation cycle, longer reactor core exposure times result in the production of curium isotopes which become the predominant source of the neutron specific activity for the spent fuel assembly. Most of the neutron emission will result from 242Cm (163 day half life) and 244Cm (17.9 year half life). For decay times more than a few years, 244Cm will be the major source of the neutron specific activity of a spent fuel assembly.
For shorter decay times, the neutron activity of 242Cm must be taken into account. Although the functional form generally covers all fuel assemblies of a particular design with different exponents for different design types, the curve will shift with initial enrichment. Therefore, knowledge of both the initial enrichment and time since discharge (decay time) are needed to accurately relate the observed neutron emission rate to burnup.
Typically, a combination of neutron measurements and reactor records are used to determine fuel burnup. In some cases, gamma ray measurements of fission product isotope gamma rays (primarily 137Cs) are used as a check on decay time. Either the gross gamma ray decay rate divided by the neutron emission rate can be related to groups of assemblies with common discharge times, or the 134Cs to 137Cs gamma decay rate ratio is measured directly to determine decay time. 134Cs has a half life of 2.06 years, and 137Cs has a half life of 30.1 years, so the decay rate ratio will change rapidly with time over a zero to 20 year time period after discharge of the fuel assembly from the reactor. A measurement of either the 134Cs/137Cs gamma emission ratio or the gross gamma emission rate is needed to verify the time since discharge (decay time) for the assembly. In the 134Cs/137Cs case, the gamma ray intensity ratio provides a direct measure of the decay time. In the gross-gamma case, the assumption is made that most of the observed activity is 137Cs, and the gross gamma to neutron ratios allow the assemblies to be separated into groups according to common discharge times. The exact decay time is then determined from fuel assembly records.
Whether in support of pool storage or loading for dry storage, the measurements are conventionally carried out under water on isolated fuel assemblies which are raised from the fuel storage rack with an overhead crane. The detector fixture is designed to attach reproducibly to the fuel assembly, and normally measurements are carried out at the fuel center line with simultaneous measurements taken on opposite fuel flats to correct for asymmetries in the neutron emission rates.
The burnup verification methodology generally requires measurements to be performed on a set of fuel assemblies of a given type. The functional form for the neutron response as a function of burnup is established on the basis of at least three measurements and is updated as data from newly measured assemblies are added. Outliers are identified on the basis of agreement with the predictions of the fit function (usually, greater than three standard deviations from the predicted value is grounds for rejection), and identified for further study. Either incorrect records or a problem with the measurement could be responsible for outlier data points.
A common feature of all of the systems presently in use is that dependence is placed on a single measurement at one axial location. The neutron emission rate at that location depends on the average axial power profile for the reactor. While some of the presently available systems are capable of measurements at multiple axial locations, a sequence of adjustments of the relative position of the detectors and assembly is required for each measurement at each axial location. This measurement process entails measurement times roughly proportional to the desired number of axial positions and greater risk of fuel damage due to the large number of movements involved.
U.S. Pat. No. 5,969,359, assigned to the assignee of the present invention, proposed an improved method and apparatus for performing measurements related to spent nuclear fuel burnup, employing miniature, temperature- and radiation-resistant semiconductor detectors that permit simultaneous monitoring of the gamma and neutron emission rates from spent nuclear fuel. Arrays of semiconductor detectors may be used to obtain information from key axial locations to define the axial burnup profile for spent nuclear fuel assemblies. Use of the improved spent fuel monitoring apparatus results in significant reductions in cost and time for the measurements, as well as improved accuracy, safety and reduced radiation doses to personnel involved in spent nuclear fuel measurements.
The nuclear detectors described in U.S. Pat. No. 5,969,359 preferably use a wide band gap semiconductor material such as SiC which is capable of providing data at elevated temperatures and is also a rad-hard semiconductor. Charged particle, neutron, and gamma ray detection may be accomplished with high-quality, miniature detectors having extremely low leakage currents which provide high-quality nuclear detection signals.
Such miniature semiconductor detectors are capable of determining neutron and gross gamma emission rates in a single measurement. A single semiconductor detector can therefore perform the functions of both the neutron and gamma detectors in the presently used systems. The preferred nuclear detectors based on SiC semiconductors are thus capable of measuring gamma rays and neutrons simultaneously in a single, energy-resolved spectrum.
Arrays of the semiconductor detectors can simultaneously measure neutron emission rates at key axial locations in order to define the shape of the fuel burnup profile. For example, a stringer of miniature semiconductor neutron detectors which simultaneously record data and are multiplexed to provide independent data for each axial location may provide information on the entire axial burnup profile. The semiconductor detector arrays can perform measurements within channels or on opposite flats of spent fuel assemblies.
The present burnup verification system using semiconductor nuclear detectors offers several advantages. For example, miniature semiconductor detectors can be positioned more accurately than the detectors presently in use, and can be used to determine fuel axial burnup profiles during a single measurement interval. While conventional methods rely on the use of an overhead crane to isolate a spent fuel assembly, the present semiconductor detectors can perform measurements on spent fuel assemblies that are located in fuel racks. In accordance with the present invention, measurements can be safely carried out with less personnel in a shorter time while eliminating the requirement of moving fuel assemblies for measurement. For instance, a single miniature semiconductor detector can provide the same gamma and neutron information as the two much larger neutron and gamma detectors used on present systems. The preferred semiconductor detectors are capable of highly stable operation in severe radiation and temperature environments. Because of these advantages, measurements made with the present semiconductor-based system are less costly than those made with commercial systems.
The present invention has been developed in view of the foregoing to further improve the burnup credit calculation to further reduce the safety margin that has to be built into spent fuel storage and shipping containers and thus further reduce the cost. While U.S. Pat. No. 5,969,359 provides a substantially improved way of measuring the axial profile, it uses that profile to more accurately calculate the total burnup of the assembly. The total burnup number is used to establish the burnup credit, however the shape of the axial burnup profile is not employed to further enhance the burn up credit. It is an object of this invention to take advantage of the axial burnup shape to further enhance the burnup credit.